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The vapor pressures of the fullerenes C60 and C70 have been measured over the temperature range 400–600 °C by the Knudsen-effusion thermogravimetric technique. For C60, a heat of sublimation of 38±1 kcal/mol is obtained, and the value for C70 is 45±1 kcal/mol. The vapor pressure of C60 ranges from 1.8×10−5 and 1.4×10−2 Torr and that of C70 is between 1.4×10−5 and 8.7×10−3 Torr over the temperature range investigated. At 500 °C, the vapor pressure of C60 is about 1035 that of graphite. The entropy of vaporization of C60 obeys Trouton’s rule.

Executive SummaryThe External Flowsheet Review Team (EFRT) expressed concern about the potential for Waste Treatment and Immobilization Plant (WTP) pipe plugging. Per the review's executive summary, "Piping that transports slurries will plug unless it is properly designed to minimize this risk. This design approach has not been followed consistently, which will lead to frequent shutdowns due to line plugging." To evaluate the potential for plugging, critical-velocity tests were performed on several physical simulants to determine if the design approach is conservative. Critical velocity is defined as the point where particles begin to deposit to form a moving bed of particles on the bottom of a straight horizontal pipe during slurry transport operations. The critical velocity depends on the physical properties of the particles, fluid, and system geometry.This report gives the results from critical-velocity testing and provides an indication of slurry stability as a function of fluid rheological properties and transport conditions that are typical of what the plant will see. The experimental results are compared to the WTP design guide on slurry-transport velocity in an effort to confirm minimum waste-velocity and flushing-velocity requirements as established by calculations and critical-velocity correlations in the design guide. The major findings of this testing are as follows:Experimental results indicate that for Newtonian fluids, the design guide is conservative. The design guide is based on the Oroskar and Turian (1980) correlation, a traditional industry-derived equation that focuses on particles larger than 100 m in size. Slurry viscosity has a greater effect on particles with a larger surface area to mass ratio, i.e. smaller particles. The increased viscous forces on small particles result in a smaller critical velocities. Since the Hanford slurry particles generally have large surface area to mass ratios, the reliance on such equations in the 24590-WTP-GPG-M-0058, Rev 0 design guide (Hall 2006) is conservative. Additionally, the use of the 95% percentile particle size as an input to this equation is conservative. The design guide specifies the use of the d 95 density, this term is ambiguous and needs clarification in the design guide. Nonetheless, this value is interpreted to mean the density of the d 95 particle. This density value is irrelevant for critical velocity calculations. Often this value is unknown, and Equation 1 of the 24590-WTP-GPG-M-0058, Rev 0 design guide (Hall 2006) will be used for design purposes. This equation calculates an average or composite density of all solids in the slurry. However, test results indicate that the use of an average particle density as an input to the equation is not conservative. Particle density has a large influence on the overall critical-velocity result returned by the correlation. The viscosity correlation used in the WTP design guide has been shown to be inaccurate for Hanford waste feed materials. Additionally, the recommendation of a 30% minimum margi...

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Summary *This report provides the results of studies conducted on coatings discovered on the surfaces of some N-Reactor spent nuclear fuel (SNF) elements stored at the Hanford K-East Basin. These elements had been removed from the canisters and visually examined in-basin during FY 1996 as part of a series of characterization tests. The characterization tests are being performed to support the Integrated Process Strategy developed to package, dry, transport, and store the SNF in an interim storage facility on the Hanford Site.Samples of coating materials were removed from K-East canister elements 2350E and 2540E, which had been sent, along with nine other elements, to the Postirradiation Testing Laboratory (327 Building) for further characterization following the in-basin examinations. These coating samples were evaluated by Pacific Northwest National Laboratory using various analytical methods. This report is part of the overall studies to determine the drying behavior of corrosion products associated with the K-Basin fuel elements.Altogether, five samples of coating materials were analyzed. These samples were identified by X-ray diffraction analysis to primarily be composed of uranium oxides and oxyhydrates. Scanning electron microscope analysis showed these samples to consist of small needles or agglomerates composed of smaller particulates and needles. This composition indicates the coatings may be formed as part of a nucleation and precipitation process. Thermogravimetric analysis combined with the total weight of material recovered from some of the elements yielded a water-content-per-surface-area-of-fuel estimate of 6-1 O4 mol water/cm2. These analyses suggest that hydration of the coating materials could be an additional source of moisture in the Multi-Canister Overpacks being used to contain the fuel for storage.iii E Acknowledgment

SummaryThe safety and process analyses of the proposed Integrated Process Strategy (IPS) to move the N-Reactor spent nuclear fuel (SNF) stored at K-Basin to an interim storage facility require information about the oxidation behavior of the metallic uranium. Limited experiments have been performed on the oxidation reaction of SNF samples taken fiom an N-Reactor outer fuel element in various atmospheres. This report discusses studies on the oxidation behavior of SNF using two independent experimental systems.. 0 a tube furnace with a flowing gas mixture of 2% oxygen/98% argon 0 a thermogravimetric system for dry air oxidation.Results of the tube furnace tests show differences in the oxidation rate of samples taken fiom the damagedhorroded and undamaged regions of the same SNF element. Samples taken from the damaged/ corroded regions of the element showed an oxygen pickup rate that was time independent, indicating a linear oxidation behavior. However, the oxygen pickup by the samples taken fiom the undamaged regions decreased with increasing time that suggests a protective oxide f i l m formation on the samples.The average oxidation rate for the undamaged SNF samples in an oxygerdargon atmosphere at 250°C was 7.01 x 10" mg/cm2/min. This rate is in close agreement with the value 4.16 x l o 3 mg/cm2/min estimated from the unirradiated data of Baker and Less and reported by Ritchie (1981).(') On the other hand, the average oxidation rate of the damaged SNF samples at 25OOC (Furnace Runs 2 1 , 23 , 27) was 2.15 x 1 0-2 mg/cm2/min, which is about a factor of 5 greater than the unirradiated uranium data of Baker and Less (Ritchie 1981):') In one particular experiment, the oxidation rate of damaged SNF sample at 150°C was 1.24 x 1 0-2 mg/cm2/min, which is about 83 times greater than the value of 1.5 x 1 O4 mg/cm2/min estimated fiom the data of Baker and Less (Ritchie 1981).(*) This sample showed significantly higher hydrogen release during its drying step(a) compared to other samples. Thus, the increased oxidation rate observed for this sample may be due to an enhancement effect of uranium hydride inclusions in the sample.The data for dry air oxidation of damagedhorroded SNF samples using the thermogravimetric (TGA/DSC/MS) system showed an oxidation behavior that was complex. The data were grouped into two categories for analysis. The first group of data represents the cases when the weight change by the SNF sample followed a typical metallic material oxidation kinetics. The second group of weight change showed oxidation behavior that was very erratic due to effects that include but are not limited to 1) sample cracking, 2) rapid oxidation of the sample resulting in large volume changes, 3) spallation of the oxide film, and 4) disintegration of the sample due to rapid oxidation of particulates in the matrix. iii The shape of the weight change data for the first group could be fitted with either linear or parabolic oxidation kinetics. A linear rate equation was fitted to the data to generate a rate equation that can be co...

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SummaryA series of tests were conducted by Pacific Northwest National Laboratory to evaluate the drying behavior of sludge taken fiom the Hanford K-East Basin storage canisters. Some of the components of K-Basin sludge, such as oxides of uranium and its hydrates, could be associated with the spent nuclear fuel that will ultimately be loaded into Multi-Canister Overpacks (MCOs) and transferred to interim dry storage on the Hanford Site. The materials sealed in the MCOs must be compatible with the storage facility safety basis and the design accident analyses. Understanding the drying behavior of hydrates that may be formed by the reaction of uranium oxides (corrosion products) and water will help ensure these criteria are addressed.Drying measurements of sludge samples collected fiom K-East Basin canisters showed the water content (physically@ plus chemically@) bound) to range between 5 wt% and 75 wt%. Uranium oxide hydrates, the main source of gaseous products that can pressurize the MCOs during storage, constituted about 3 wt% to 15 wt% of the total water content of the initial weight. Most of the physically bound water was assumed to be released from the samples at ambient temperature when the system was pumped down to vacuum conditions of about 40 mTorr. The period for release of most free water in the K-East canister sludge was about 24 hours.Fourteen tests were run on subsamples from a batch of six larger K-East canister samples. The test temperatures ranged from ambient to about 625°C. For the last test, in which the drying was conducted in helium at atmospheric pressure, a partial loss of the physically bound water was observed before the start of the test. The complete release of the large fraction of free water remaining in the sample did not occur until the sample was heated to 75°C. This suggests that when a reasonable quantity of fiee water was present in the sample, the effective removal of the free water from the sludge occurred at temperatures equal to 75 "C and above. This observation supports the assertion that most of the weight losses by the samples at temperatures below 75°C were caused by desorption of the remaining free water rather than by thermal decomposition of hydrates.The thermal decomposition of the hydrates effectively started when the sample temperature was about 75 "C. From the results of the test runs and analyses, thermal decomposition of the hydrates appears to drive the release of water, resulting in sample weight loss at temperatures between 75 "C and (a) Physically bound water is the "free water" that is physically absorbed by the pores and adsorbed by (b) Chemically bound water is the hydrate(s) and hydroxides formed by the reaction of water with the the surface of the sludge.sludge; referred to here as "hydrates" or "hydrated species." iii 425°C. At temperatures above 425"C, the weight losses observed were due to reduction of higher oxides of uranium by reactions such as rather than decomposition of hydrates.The measured weight losses for ,he test temperature segments are ...